ML20216A905
| ML20216A905 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 05/06/1998 |
| From: | Johnson W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20216A877 | List: |
| References | |
| 50-482-98-10, NUDOCS 9805140295 | |
| Download: ML20216A905 (27) | |
See also: IR 05000482/1998010
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ENCLOSURE 2
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No.: 50-482
License No.: NPF-42
Report No.: 50-482/98-10-
Licensee: Wolf Creek Nuclear Operating Corporation
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Facility: Wolf Creek Generating Station
Location: 1550 Oxen Lane, NE
Burlington, Kansas
Dates: March 8 through April 18,1998
Inspectors: J. F. Ringwald, Senior Resident inspector
B. A. Smalldridge, Resident inspector
R. V. Azua, Project Engineer
K. M. Thomas, Project Manager, Office of Nuclear Reactor Regulation
Approved: W. D. Johnson, Chief, Project Branch B
ATTACHMENT: Supplemental Information
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9805140295 990506 2
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EXECUTIVE SUMMARY
Wolf Creek Generating Station
NRC Inspection Report 50-482/98-10
Ooerations
. A shift supervisor made an appropriate, but inadequately' documented operability
determination associated with abnormal noises from an essential service water traveling
screen (Section O4.1)
. Operations training personnel conducted effective simulator training and the staff crew
responded to the simulated steam generator tube rupture in an effective manner.
Training personnel initiated performance improvement requests (PIRs) to address the
fact that they lacked documented standards as to what constituted a complex scenario
and for what types of operator errors would require corrective action beyond that
afforded by the postscenario critique (Section 05.1).
. Licensee personnel had not monitored telephone calls to the quality first hotline between
September 13,1997, and April 1,1998 (Section O6.1).
. An operations self-assessment on the use of the operations evolution checklist identified
that the checklist was not being used consistently and in its entirety each time, yet did
not identify this as a problem. The self-assessment team's recommended actions to
. reduce the scope and level of detail of the checklist were implemented, making the
checklist less effective (Section O7.1).
. ' The licensee identified several violations of NRC requirements and reported these in
licensee event reports (LERs). These licensee-identified and corrected violations are
being treated as noncited violations, consistent with Section Vll.B.1 of the NRC
Enforcement Policy (Section 08).
Maintenance
.- The material condition of those plant systems and components evaluated during this
inspection period was good, with few equipment deficiencies. The inspectors noted an
increase in the number and severity of boric acid leaks in the auxiliary building and that
an oil leak caused an unplanned inoperability of a centrifugal charging pump
(Section M2.1).
. An arc flash event occurred during the replacement of a molded case breaker when the
breaker was not adequately disconnected from the bus. - The corrective action failed to
identify that the clearance order was incomplete in that the clearance did not establish a
boundary to isolate the breaker from the bus (Section M3.1).
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Engineering
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The inspectors found that the review level of the work package and the procedure used
to transfer an irradiated specimen was not commensurate with the potential risk to
workers involved in the transfer process. No procedure review beyond that of the
engineers working on the project was accomplished prior to the inspectors' observations
(Section E1.1).
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Based on a review of select changes made to the Updated Safety Analysis Report,
Revision 11, the project manager determined that the changes at Wolf Creek Generating
Station have been appropriately addressed and that the changes to the Updated Safety
Analysis Report have been appropriately addressed by licensing actions,10 CFR 50.59
submittals, etc. (Section E2.1).
Plant Suonort
. Radiation protection technicians improperly posted an area with dose rates of
approximately 400 mrem /hr at 18 in. After discovery, the technicians properly posted
but failed to barricade the area (Section R1.1).
Security personnelinadvertently issued a locked high radiation area key to a radwaste
operator who was not authorized to receive the key. This licensee-identified and
corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1
of the NRC Enforcement Policy (Section R1.2).
. Two contaminated area postings did not clearly describe the location of contamination on
valves (Section R1.3).
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The inspectors found the level of the as-tow-as-reasonably-achievable (ALARA) review
cf the irradiated specimen transfer work package and procedure not commensurate with
the potential risk to workers involved in the specimen transfer process. The inspectors
determined that this was because the licensee's health physics program did not include
the potential for high exposure in the criteria for determining when an ALARA review
should be performed (Section R3.1).
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Reoort Details
Summarv of Plant Status
The plant operated at essentially 100 percent power throughout the inspection period.
I. Operations
04 Operator Knowledge and Performance
04.1 Inadeauntelv Documented Ooerability Evaluation
a. Insoection Scona (711QZ)
The inspectors reviewed one documented Technical Specification operability screening
checklist during the inspection period.
b. Observations and Findinos
On March 6,1998, the shift supervisor completed a Technical Specification operability
screening checklist to document an operability determination performed on the Essential
Service Water Train B traveling screen. During operation, abnormal noises were heard
from the traveling screen. Maintenance personnel and the system engineer evaluated
the noise and engaged in several troubleshooting activities. Afterwards, the system
engineer stated that the noise did not affect the operability 2f the traveling screen, but
provided no basis for this conclusion. The shift supervisor completed the operability
checklist and asserted that the essential service water train was operable. The basis
statement in the operability determination described the noise and troubleshooting
activities and provided the system engineer's judgement. However, the basis statement
did not provide a tangible reason for why the noise did not represent, and could not lead
to, inoperability of the traveling screen.
After the inspector raised the question, the operations manager directed the shift
supervisor to amend the operability checklist to provide a basis. The additional
information provided an adequate basis for the operability of the traveling screen.
Operations personnel subsequently initiated PIR 98-0699 to address the recurring
concern of shift supervisors making appropriate but inadequately documented operability
determinations.
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c. Conclusions
A shift supervisor made an appropriate but inadequately supported operability
determination associated with abnormal noises from an essential service water traveling
screen.
05 Operator Training and Qualification
05.1 Simulator Trainina
a. insoection Scone (71707)
The inspectors observed a complex simulator training scenario. ,
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b. Observations and Findinos
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On April 2,1998, the inspectors observed a complex simulator training scenario for an '
operations staff crew. The scenario involved a steam generator tube rupture that
exceeded the available makeup rate. Operators appropriately recognized the symptoms I
and diagnosed the problem. Their response was generally appropriate. The instructor i
effectively controlled the scenario to ensure that appropriate training objectives were
accomplished.
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The training personnel considered this scenario complex because it presented operators (
with a number of challenges, including a failure of the loss-of coolant accident J
sequencer, a stuck closed reactor trip breaker, the inability to restore instrument air to
containment, and the failure of a boron injection tank flow indicator. These challenges
occurred in a relatively linear manner and, therefore, did not present the supervising l
operator with a significant prioritization challenge. As a result, from a command and i
control standpoint, training provided a scenario that was not appreciably complex. The {
inspector asked how this scenario met the criteria to be classified as a complex scenario. !
The operations training superintendent responded by initiating PIR 98-1249 to note and
correct the current absence of documented criteria for what constitutes a complex )
scenario. d
The instructor noted a significant emergency management guidelines procedure usage
error in that the supervising operator failed to refer to the response-not-obtained section
of Procedure EMG E-0, " Reactor Trip or Safety injection," Revision 11, associated with '
Step 1.b, to ensure that reactor trip breakers and bypass breakers were open. The.
inspector asked the operations training superintendent what corrective actions had been
taken in response to the identification of this error.. The superintendent responded that,
in this case, the instructor considered the discussion of the error during the critique to be
adequate. However, since the operations training superintendent did not have a
documented criteria to address what type of operator errors identified during training
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would most appropriately be addressed solely by the critique, and when additional
actions would be needed, the superintendent initiated PIR 98-1249 to address the
absence of this criteria.
While the instructor noted several critique issues and effectively addressed them with the
crew during the critique, the inspectors identified two errors which were not discussed by
either the instructor or the crew during the critique. First, operators acknowledged
several annunciators prior to the reactor trip without referencing the associated alarm
response procedure. Second, the supervising operator reported that the leak rate
decreased when instruments indicated that the leak rate increased. The shift supervisor
apparently recognized and mentally corrected this erroneous report without specifically
discussing the matter with the supervising operator, as all operators responded to the
increasing leak rate appropriately. However, neither the operators nor the instructors
noted this communication error. .The operations training superintendent informed the
inspector that the management expectations were for instructors to note and discuss all
significant operator errors during the critiques immediately following simulator training
scenarios. The superintendent stated that this expectation will be reinforced with all
simulator instructors.
Despite the fact that the observed crew was a staP crew, the operators demonstrated
generally effective use of three-way communication. In some instances the instructor
noted communications that were not three-way; however, these were infrequent. The
inspector noted that, early in the scenario, communications were not consistently
three-way, but as the scenario progressed, the use of three-way communication became 1
more consistent,
c. Conclusions
Operations training personnel conducted effective simulator training and the staff crew
responded to the simulated steam generator tube rupture in an effective manner.
Training personnel initiated PIRs to address the fact that they lacked documented
standards for what constituted a complex scenario and for what types of operator errors
would corrective actions, beyond that afforded by the postscenario critique, be required.
06 Operations Organization and Administration
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06.1 Quality First Teleohone System ineffective
a. Insoection Scone (71707) l
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The inspector evaluated the effectiveness of the quality first telephone system.
b. Observations and Findinas
On March 2,1998, the inspector telephoned the quality first telephone number and
reached the voice mail box for the quality first program. The inspector left a message
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asking a quality first representative to retum the telephone call. On April 1,1998, the
inspector noted that no retum telephone call had been made and asked wny this
occurred.
After researching the issue, licensee personnel informed the inspector that, due to a
miscommunication, the quality flrst voice mail box had not been effectively monitored
since September 13,1997. The responsibility for monitoring the quality first voice mail
box changed in September 1997, and the new person failed to receive or understand the
instructions. Consequently, the messages were not being retrieved when the person
believed that they were.
After discovering this problem, licensee personnel retrieved all messages from the voice
mail box. Only two messages had been left in the quality first voice mail box, the one
from the inspector and one from a licensee employee who asked an administrative
question but did not raise a nuclear safety concern. The licensee initiated PIR 98-0936
to address this problem and establish long-term corrective actions.
c. Conclusions
Licensee personnel had not monitored telephone calls to the quality first hotline between
September 13,1997, and April 1,1998.
07 Quality Assurance in Operations
O7.1 Ooerations Self-Assessment Not Sufficientiv Critical
a. Insoection Scooe (71707)
The inspector reviewed Self-Assessment Report SEL 98-003, " Effectiveness of the
Operations Evolution Checklist."
b. Observations and Findinas
This self-assessment reviewed the effectiveness of the operations evolution checklist, a
checklist that the licensee described as part of their corrective actions for the event
described in LER 50-482/96-008. The commitment was proceduralized in Administrative
- Procedure AP 21-001, " Operations Watchstanding Practices," Revision 6.
The self-assessment results section described interview results involving operations and
integrated plant scheduling personnel who were required by licensee procedures to use
the checklist every time certain evolutions were to begin. During these interviews, only
one person said that they used the checklist in its entirety every time, and several
individuals said that, when the workload was heavy, the use of the checklist in its entirety
each time decreased.
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The inspector questioned whether this statement in the self-assessment report
suggested that licensee personnel were not complying with the procedural requirements.
Operations personnel responded that this was not what was meant by the words in the
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self-assessment report. Instead, the interviewers explained that what they understood
the responses to mean was that certain steps of the checklist were not considered
applicable to certain tasks and, therefore, that that checklist element was not considered
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The inspector also noted that the self-assessment report recommended a revision to the 1
checklist that made it shorter and asked questions of a more general nature than the
s original checklist. The inspector determined that most cognizant individuals would be
able to envision a majority of the' elements of the proposed checklist, without the -l
prompting provided by the checklist. As a result, the revised checklist appeared to '
provide less assistance to operations than the previous checklist.
On March 20,1998, the licensee issued Administrative Procedure AP 21-001,
Revision 9, to implement the revised checklist. The shorter, less detailed checklist
recommended in the self-assessment report was implemented as recommended.
After the inspector raised questions with the self-assessment, operations personnel
initiated PIR 98-0983 to address the problem identified in the self-assessment report.
c. Conclusions
An operations self-assessment on the use of the operations evolution checklist identified
that the checklist was not being used consistently and in its entirety each time, yet did
not identify this as a problem. The self-assessment team's recommended actions to
reduce the scope and level of detail of the checklist were implemented, making the
checklist less effective.
08 Miscellaneous Operations issues (92901)
08.1 - (Closed) LER 50-482/9618-00: Failure to comply with Technical Specification
Surveillance Requirement 4.3.2.2. This item was identified when the licensee reviewed
the auxiliary feedwater system's licensing basis documentation as part of an overall
functional assessment of the auxiliary feedwater system. Technical Specification
- Requirement 4.3.2.2 for engineered safety features response-time testing required that it
be verified that the turbine-driven auxiliary feedwater pump be able to provide flow to the
steam generators within 60 seconds of an initiating signal by reaching its rated speed of
3850 rpm. The licensee noted that Surveillance Procedure STS AL-104, " Turbine Driven
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Auxiliary Feedwater Pump Engineered Safety Features Response Time Test," verified
the turbine-driven auxiliary feedwater pump actuation response time indirectly by
measuring the turbine throttle valve stroke open time as opposed to monitoring the pump
speed directly. Two previous opportunities were identified where the licensee failed to !
identify this error. Failure by licensee management to ensure that there were proper
mechanisms in place to verify that their standards and expectations were being met was
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determined to be the root cause. Licensee corrective actions to mis event included the !
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development of a corrective action review board, changing the leadership of the plant
safety review committee, revising the surveillance procedure, and performing a sample
review of other serveillances. The described corrective actions were found to be
appropriate to address this issue. The failure by the licensee to ensure that surveillance
procedure guidance was consistent with Technical Specification requirements is a
violation of NRC requirements. This licensee-identified and corrected violation is being
treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement
Policy (50-482/9810-01).
08.2 (Closed) LER 50-482/9620-00: Failure to comply with Technical Specification Action
. Statement 3.6.1.7.a. This item was identified when the licensee reviewed the draft
improved Technical Specification 3.6.1.7, " Containment Ventilation Systems," as part of
a Technical Specification improvement process, and recognized that the current
Technical Specification action statement allowed only one 3F anch containment purge
supply and/or isolation valve to be opened or not blank flanged at any one time.
Surveillance Test Procedure STS PE-015, "18-inch Purge Valve Leakage Test," which
satisfied Surveillance Requirement 4.6.1.7.4, was found to conflict with the Technical
Specification because it required the opening of both 36-inch containment purge supply
valves simultaneously. This item was identified because of the licensee's efforts to
enforce strict literal compliance with Technical Specifications. The root cause of this
. event was determined to be a misalignment between the organizational culture and the
regulatory environment.. Corrective actions taken included revising Procedure STS PE-
015, reviewing other surveillance procedures related to Technical Specification 3.6.1.7, .
providing training sessions with all departments regarding literal compliance with NRC '
requirements, and providing training to insure the proper alignment exists between the
Wolf Creek culture and the regulatory environment. The described corrective actions
were found to be appropriate for addressing this issue. The failure by the licensee to
ensure that surveillance procedure guidance was consistent with Technical Specification
requirements is a violation of NRC requirements. This licensee-identified and corrected
violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the
NRC Enforcement Policy (50-482/9810-02).
08.3 (Closed) LER 50-482/9701-00. 01. 02. 03: Failure to comply with the surveillance
requirement to test certain components at shutdown. This item was identified by the-
licensee during a quality evaluations audit, during which it was questioned whether the
surveillance requirements for Technical Specification 4.7.3.b were being appropriately
met. Technical Specification 4.7.3.b stated in part that "at least two component cooling
water loops shall be demonstrated operable once per 18 months during shutdown." It
was found that portions of the Technical Specification surveillance required to be. ,
performed at shutdown were performed in Mode 1. The surveillance in question was i
performed within the appropriate time constraints set forth in the Technical l
Specifications, and the surveillance results met the acceptance criteria. A review of I
other Technical Specifications by the licensee resulted in the identification of 24 other
occurrences where surveillance tests were performed at modes other than at shutdown
contrary to Technical Specification requirements. No significant operational concerns
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were identified. The root cause of these events was determined to be personnel errors
made in scheduling of Technical Specification surveillances. The licensee's corrective
actions included a review of the scheduling database for similar scheduling errors,
revision of applicable procedures, and reinforcement of management expectations with
the current operations procedure writers and schedulers. The described corrective
actions were found to be appropriate for addressing this issue. The failure by the
licensee to ensure that surveillance tests were performed during the appropriate modes
of operation as required by Technical Specifications is a violation of NRC requirerrs.ts.
This licensee-identified and corrected violation is being treated as a noncited violation,
consistent with Section Vll.B.1 of the NRC Enforcement Policy (50-482/9810-03).
08.4 (Closed) LER 50-482/9704-00: Failure to accurately perform the containment integrated
leak rate test. This item was identified by the licensee. Technical Specification 4.6.1.1.d
stated that primary containment integrity shall be demonstrated by performing
containment leakage rate testing in accordance with the containment leakage rate testing
program. The licensee's Updated Safety Analysis Report indicated that, to assure leak
tight integrity, the hydrogen analyzer containment penetration valves would have to be
subjected to Type C testing and the sample lines would be opened during the Type A
test. The last Type A test, which was performed in 1991, was reviewed and it was
determined that these containment penetration valves associated with the hydrogen
analyzers were closed during the Type A test. Procedure CKL PE-018, "lLRT System
Lineups," provided erroneous guidance. The root cause of this event was determined to
be an incorrect startup procedure which was developed as part of the containment
leakage testing program. Licensee corrective actions included revising applicable
procedures, performing the required leak rate test to verify compliance with 10 CFR
Part 50, Appendix J, requirements, training of personnel, and performance of a
self-assessment of the containment integrated leak rate program. The described
corrective actions were found to be appropriate for addressing this issue. The failure by
the licensee to ensure that surveiliance procedure guidance was consistent with
Technical Specification requirements is a violation of NRC requirements. This licensee-
identified and corrected violation is being treated as a noncited violation, consistent with
Section Vll.B.1 of the NRC Enforcement Policy (50-482/9810-04).
08.5 (Closed) LER 50-482/9705-00: Violation of Technical Specification 3.8.1.1.b Emergency
Diesel Generator B. Licensee personnel, while working on an unrelated activity,
identified that the screws needed to secure the back panel of the Emergency Diesel
Generator B annunciator assembly were missing. The licensee determined that the back
panel could potentially cause the inoperability of Emergency Diesel Generator B under
seismic conditions. The root cause of this event could not be determined exactly, but
human error was considered a factor. The licensee fastened the back panel for the
Emergency Diesel Generator 8 annunciator assembly, inspected the annunciator back
panel for the Emergency Diesel Generator A, reviewed previously performed work
packages in an attempt to determine when this error occurred (no activities were
identified that would have called for the removal of the panel in question), and provided
training to maintenance craft and planning personnel to reinforce the expectations of job
completion. The described corrective actions were found to be appropriate for
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addressing this issue. The failure by the licensee to ensure that all conditions were
appropriate for maintaining the operability of the emergency diesel generators as
required by Technical Specifications is a violation of NRC requirements. This licensee-
identified and corrected violation is being treated as a noncited violation, consistent with
Section Vll.B.1 of the NRC Enforcement Policy (50-482/9810-05). !
08.6 (Closed) LER 50-482/9707-00. 01: Failure to procure test gas for a safety-related
component through the 10 CFR Part 50, Appendix B, program. The licensee identified
this issue while verifying prerequisite requirements for performance of Surveillance Test
Procedure STS IC-913, "CTMT Hydrogen Analyzer GS0658 Cal Test." Instrumentation
and controls technicians noticed discrepancies on the test certificates located with the
Hydrogen Analyzer B calibration test gas bottle. The licensee reviewed the appropriate
- documentation and was unable to certify the concentration of the contents of the
calibration gas bottle. Further review identified that the calibration test gas was not
purchased through the 10 CFR Part 50, Appendix B, program because originally during
plant startup the test gas was considered nonsafety-related. The root cause of this event
was determined to be personnel error la f#ng to follow Pro :edures AP 24-002,
" Requisition, Procurement Process," a9] AP 24E-001, " Identification and Control of
Materials, Parts, and Components." Lack of reconciliation of nonsafety-related material
for use in calibrating a safety-related piece of equipment has been determined to be a
contributing factor. Corrective actions taken included:
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Performing a safety classification analysis for the hydrogen analyzer test gas
which determined that the certification process for the test gas was safety-related,
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Performing an inspection on the vendor of the calibration gas and the licensee
issued a commercial grade dedication,
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Purchasing safety-related test pas for the containment hydrogen analyzer,
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Performing a review of other test gasses to determine if they had been
appropriately purchased (no problems were noted), and
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Providing training to maintenance personnel involved in originating material
requisitions.
The described corrective actions were found to be appropriate for addressing this issue.
Licensee failure to follow approved procedures in purchasing material for testing
safety-related equipment is a violation of NRC requirements. This licensee-identified and
corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1
of the NRC Enforcement Policy (50-482/9810-06).
08.7 (Closed) LER 50-482/9709-01: Failure to comply with Technical Specification 4.5.2.c.2.
This event was discussed in NRC Inspection Report 50-482/97-10. No new issues were
revealed by the LER.
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08.8 (Closed) LER 50-482/9713-00: Section 3.7.3 Technical Specification violation regarding
two independent component cooling water loops. Licensee personnel identified that the
component cooling water loops had been cross-tied on five different occasion *, when
Procedure SYS fig-202, "CCW Surge Tank Level Equalization," was used. The
procedure guidance was found to be in error. The root cause of this event was found to
be human error in that licensee personnel failed to recognize and apply the requirem',nts
of Technical Specification 3.7.3 (which required "At least two independent component
cooling water loops shall be operable" during Modes 1 through 4) during revisims to
operating procedures which allowed cross-connecting the comoonent cooling water
surge tanks. The licensee's corrective actions included canceling Procedure
SYS EG-202 and reviewing the SYS series of operating procedures. The described
corrective actions were found to be appropriate for addressing this issue. The failure by
the licensee to ensure that surveillance procedure guidance was consistent with
Technical Specification requirements is a violation of NRC requirements. This licensee-
identified and corrected violation is being treated as a noncited violation, consistent with
Section Vll.B.1 of the NRC Enforcement Policy (50-482/9810-07).
08.9 (Closed) LER 50-482/9714-00: Waste gas system hydrogen / oxygen analyzers not
operated in accordance with Technical Specification requirements. The licensee
identified this issue when a radwaste operator, while performing a daily surveillance ;
proceaure, questioned the acceptability of operating a waste gas system compressor j
with both hydrogen / oxygen analyzer racks being out of service. Technical l
Specifications 16.3.1.6 and 16.11.2.1.1 require that the hydrogen / oxygen analyzers be in
service at all times during waste gas holdup system operation. The root cause of this
event was determined to be inadequate administrative controls for ensuring that
analyzers were in service during system operation. Contributing factors included a lack
of clear definition of what constituted the waste gas system being in service. Licensee
corrective actions included placing the waste gas analyzers in service and revising
selected procedures to ensure that waste gas analyzers are placed in service when the
waste gas compressor is operating. The described corrective actions were found to be
appropriate for addressing this issue. The failure by the licensee to ensure that
surveillance procedure guidance was consistent with Technical Specification
requirements is a violation of NRC requirements. This licensee-identified and corrected
violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the
NRC Enfcrcement Policy (50-482/9810-08).
08.10 (Closed) LER 50-482/9719-00: Failure to adequately vent emergency core cooling pump
casings and discharge piping high points in accordance with Technical Specification
Surveillance Requirement 4.5.2b.1. The licensee identified this issue following
discussions with another utility and after reviewing surveillance Procedure STS BG-002,
'ECCS Valve Check and System Vent." This procedure did not provide for venting it'e
centrifugal charging pump casings since the design of the pump did not include casing
vents. The residual heat removal pump casings had not t,9en vented in accordance with
the surveillance requirement. The root cause of this event was determined to be
personnel error on the part of the initial procedure writers in that the requirements of the
Technical Specifications were not understood or clarified before procedures were initially
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written or, subsequently, when they were revised. The licensee's corrective actions
included: submitting an exigent license amendment request to the NRC to revise the
surveillance requirement to require venting of only the residual heat removal and safety
injection, pump casings; reviewing isometric drawings of the residual heat removal,
safety injection, and charging systems to identify the discharge piping high point vents;
and revising Procedure STS BG-002 appropriately. The described corrective actions
were found to be appropriate for addressing this issue. The failure by the licensee to
ensure that surveillance procedure guidance was consistent with Technical Specification
requirements is a violation of NRC requirements. This licensee-identified and corrected
violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the
NRC Enforcement Policy (50-482/9810-09),
ll. Maintenance
M1 Conduct of Maintenance
M1.1 General Comments on Maintenance Activities
a. Insoection Scoce (72707)
The inspectors observed all or portions of the following work activities.
WP 101433 Task 8 NG01 A CR2 molded case breaker replacement
WP 120634 Task 1 Motor-operated valve preventive maintenance on
EJ FCV610
WP 123484 Task 1 Motor-operated valve preventive maintenance on
BN HV8812A
WP 101435 Task 14 NG02A remove existing breaker and install
Westinghouse breaker
WP 112855 Task 2 Replace Pump PAN 01A
WP 126639 Task 1 Irradiated specimen transfer
b. Observation and Findinas ;
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Except as noted in Sections M2.1 and M3.1, the inspectors found no concerns with the
maintenance observed.
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c. Conclusions
Except as noted in Sections M2.1 and 3.1, the inspectors concluded that the
maintenance activities were being performed as required.
M1.2 General Comments on Surveillance Activities
a. Insoection Scone (61726)
The inspectors observed all or portions of the following surveillance activities.
STS EG-201A, Revision 3 Component cooling water system Train A inservice valve
test
STS EN-1008, Revision 11 Containment Spray Pump B inservice pump test
STS KJ-005A, Revision 30 Manual / Auto start, synchronization loading of Emergency
Diesel Generator NE01
STS RE-015, Revision 2 Irradiated surveillance specimens
STS SE-001, Revision 21 Power range adjustment to calorimetric
b. Observations and Findinas
The inspectors found no concerns with the surveillances observed.
c. Conclusions
The inspectors concluded that the surveillance activities were being performed as
required.
M2 Maintenance and Material Condition of Facilities and Equipment
M2.1 Review of Material Condition Durina Plant Tours
a. Insoection Scooe (61726)
During this inspection period, routine plant tours were conducted to evaluate plant
material condition.
b. Observations and Findinas
in general, where equipment deficiencies existed, the deficiencies had been identified by
the licensee for corrective action.
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During this inspection period, the inspectors noted an increase in the number and
severity of boric acid leaks from valves, flanges, and components in the safety
injection pump rooms, centrifugal and normal charging pump rooms, residual
heat removal heat exchanger rooms, and the mechanical penetration rooms over
conditions that existed previously. The inspectors noted that the licensee han
taken action to correct this trend.
.
The inspectors noted that the condition of the Class 1E air conditioning unit
(SGK 05A/B) discharge valves was degrading due to worsening corrosion in the
bonnet area.
.
On March 30,1998, an oil leak on Centrifugal Charging Pump B resulted in an
unplanned entry into a limiting conditions for operation action statement due to
the pump being inoperable. The inspectors noted that the licensee restored the
system to operable status promptly and appropriately.
F
c. Conclusions
The material condition of those plant systems and components evaluated during this
inspection period was good, with few equipment deficiencies. The inspectors noted an
increase in the number and severity of boric acid leaks in the auxiliary building and that
an oilleak caused an unplanned inoperability of a centrifugal charging pump.
M3 Maintenance Procedures and Documentation
M3.1 Molded Case Breaker Arc Flash Event
a. Insoection Scoos (62703)
The inspector reviewed the licensee's response to an arc flash event that occurred
during a molded case circuit breaker replacement.
b. Observations and Findinas
On March 17,1998. during the replacement of a molded case circuit breaker in an ITE
Gould Series 5600 motor control center, the electrician removed the 24 inch
self-contained control unit and relied on the dogs to prevent the control unit from moving
inward and contacting the energized bus. While tightening the screws at the top of the
breaker, the electrician exerted sufficient force on the self-contained control unit to cause
it to flex, which closed the gap and caused the contacts to touch the bus. This created
an electrical current path from the bus to the control unit contacts, through the wires to
the line side breaker terminals, up the electrician's screwdriver shaft, to a bracket holding
the breaker, to the control uriit metal housing, to the control unit dogs, to the motor
control center frame, to ground. In the process, an arc occurred at several points along
this path, including one between the electrician's screwdriver and th~ e breaker mounting
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bracket and one which welded the control unit dog to the motor control center frame.
The screwdriver arc caused a flash burn to the electrician's finger. The electrician was
given first aid and permitted to return to work.
The licensee initiated a safety investigation which identified that the electrical isolation for
the control unit was not adequate. By relying on the dogs to prevent inward movement
of the control unit, the gap between the contacts on the back of the control unit and the
bus was approximately 1/4 inch. A:, demonstrated by this event, control unit flexure can
be large enough to close this gap and bridge the presumed electrical isolation.
The vendor technical manual procedure for locking out these control units for
maintenance described 3/8-inch holes in the motor control center slide rail and in the
control unit. With these holes aligned, the clearance between the bus and the control
unit contact increased 7/8 inch beyond the clearance provided by reliance on the dogs.
The licensee discussed this event with several other utilities and determined that every 1
utility they contacted had been isolating motor control center control units in the same )
manner and were similarly unfamiliar with the vendor technical manual guidance !
regarding the control unit lockout position. NRC Morning Report 4-98-0013 provides an
initial overview of this event. The licensee reported this event to the LJustry through the
institute for Nuclear Power Operations and is evaluating other means to r ommunicate
this information more widely.
The licensee revised the maintenance procedure for these breakers to require
electricians to slide the control unit out sufficiently so these lockout holes align and to
install a locking device that prevents inward movement.
The inspector noted that the licensee's investigation did nd specifically conclude that the
clearance order for this work was inappropriate. The inspector reviewed the clearance ,
order for this breaker replacement. The clearance order required that the breaker be l
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tagged open, but provided no boundary to isolate the molded case breaker from the bus.
When the inspector asked clearance order personnel why this clearance order did not
establish isolation of the molded case breaker from the bus, they replied that it was their
standard practice to rely on the electrician's skill of the craft and on the procedure to
isolate the breaker from the bus before de-terminating it. The inspector reviewed the
licensee's program for working on known energized equipment. The program only
applies to circuits with voltages exceeding 277 volts and only applies to work involving
planned contact of the electrician's hand or tools with known energized conductors.
Since the electricians planned to slide the control unit out prior to de-terminating the
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breaker, they did not plan to work on energized equipment and, therefore, this program
did not apply.
The inspector reviewed Procedure AP 21E-001, " Clearance Orders," Revision 6, and
noted that the procedure purpose statement read "This procedure provides a tagging
method to ensure personnel safety and equipment protection during all plant conditions."
The inspector questioned whether the clearance order procedure specifically required
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components to be replaced to be completely isolated from energy sources. The
inspector also questioned whether the clearance order procedure permitted issuance of
clearance orders that did not provide complete isolation and, therefore, protection of
personnel and equipment. Finally, the inspector questioned whether there was a
provision in the clearance order program for the preparer to intentionally not completely
isolate a component and visibly invoke some sort of other administrative controls to
- ensure that these other controls provided the protection not provided by the clearance
order. These questions were still open at the end of the inspection period, and these
issues will be tracked as an unresolved item until these questions are answered
(50-482/9810-10).
c. Conclusions
An arc flash event occurred during the replacement of a molded case breaker when the
breaker was not adequately disconnected from the bus. The corrective action failed to
identify that the clearance order was incomplete in that the clearance did not establish a
boundary to isolate the breaker from the bus.
Ill. Engineering
E1 Conduct or Engineering
E1.1 Enoineerina Daartment Review of Irradiated Soecimen Transfer Work Packaoe
a. Insoection Scooe (37551)
The inspector reviewed engineering department activities associated with transfer of an
irradiated specimen from the spent fuel pool to a shipping cask.
b. Observations and Findinas
On March 25,1998, the inspector observed engineering, operations, and health physics
personnel prepare to accomplish movement of an irradiated reactor vessel specimen
from the spent fuel pool, where it had been stored since Refueling Outage 9, to a
shipping cask along side the pool. The evolution was accomplished using Work
Package 126639, Irradiated Specimen Transfer, Task 1, and Procedure STS RE-015,
' irradiated Surveillance Specimens," Revision 2.
The irradiated specimen transfer process had been accomplished twice before this time,
most recently in December 1991. This infrequent task involved raising the irradiated 1
specimen from the spent fuel pool using the spent fuel pool crane and lowering the
specimen into a shipping cask with a narrow (just-the-right-size) opening. This process
was complicated by the fact that high radiation doses from the specimen (40R/hr)
required the operators to perform this operation from a distance using a television
monitor in order to reduce their radiation dose. Further complications were encountered
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when the gripper used to grapple the irradiated specimen in the spent fuel pool did not
match the connection points on the' specimen liner. Despite these complicating factors
and the potential for high doses should a problem have been encountered, no procedure
review beyond that of the engineers working on the project was accomplished prior to the
inspectors observations.
The inspectors found that the level of review of the work package and the precedure
used to transfer the irradiated specimen was not commensurate with the potential risk to
workers involved in the transfer process. The inspector also determined that the
- installed funnel used to guide the specimen into the shipping cask was not addressed in
the work package or in the procedure. Questions asked by the inspector resulted in a
change being made to the procedure to address use of the funnel as a guide for the
specimen.
c. Conclusions
The inspectors found that the level of review of the work package and the procedure
used to transfer an irradiated specimen was not commensurate with the potential risk to
workers involved in the transfer process. No procedure review beyond that of the
engineers working on the project was accomplished prior to the inspectors' observations.
E2 Engine *, ring Support of Facilities and Equipment .
E2.1 Review of Wolf Creek Generatina Station Annual Safety Evaluation Reoort and Final '
Safety Analvsis Reoort Uodate
a.- Insoection Scooe (37551)
By letters dated March 11,1998, Wolf Creek Nuclear Operating Corporation submitted
the Wolf Creek Generating Station Annual Safety Evaluation Report and the Wolf Creek
Updated Safety Analysis Report (USAR), Revision 11. During the week of March 23,
1998, the NRC project manager performed a review of the annual safety evaluation !
report on a sampling basis to assess whether the licensee's submittal clearly described
the changes in sufficient detail to determine if the licensee's conc;usion that the changes
did not involve an unreviewed safety question appeared reasonable. In addition, the {
NRC project manager performed a review of Revision 11 to the USAR on a sampling J
basis to determine if: (1) changes at the Wolf Creek Generating Station have been ]
appropriately addressed in the USAR, and (2) changes to the USAR have been I
appropriately addressed by licensing actions,10 CFR 50.59 submittals, etc.
b. Observations and Findinas
The project manager determined that the safety evaluation summaries in the annual
safety evaluation report that were reviewed were of high quality and contained sufficient
detail to determine if the licensee's conclusion that the changes did not involve an
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unreviewed safety question appeared reasonable. A review, on a sampling basis, of the
safety evaluations summarized in the annual safety evaluation report, will be performed
during the next 10 CFR 50.59 inspection.
. ConclusiODE
Based on a review of select changes made to the USAR, Revision 11, the project
manager determined that the changes at Wolf Creek Generating Station have been
appropriately addressed in the USAR and that the changes to the USAR have been
appropriately addressed by licensing actions,10 CFR 50.59 submittals, etc.
IV. Plant Support
R1 Radiological Protection and Chemistry Controls
R1.1 Inadeouate Hiah Radiation Area Postina
a. Insoection Scone (71750) j
The inspector reviewed the licensee's response to the discovery of an inappropriately
posted high radiation area.
b. Observations and Findinas
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On April 10,1998, a radiation protection supervisor noted that health physics technicians
posted the entrance to scaffolding up to the roof of the new radwaste storage building as !
a radiation area. The supervisor questioned whether the radwaste building roof was
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accessible from the roof of the new radwaste storage building. The survey map that I
supported the posting for the new radwaste storage building showed radiation levels at i
the edge of the new radwaste storage building had been 60 mrem per hour and also )
showed that the technicians had not documented surveying the roof of the radwaste ]
building. The subsequent investigation revealed that the technicians had not surveyed )
the roof of the radwaste building and that it would be easy to step from the new radwaste t
storage building roof to the radwaste building roof. Subsequent surveys showed that I
radiation levels on the radwaste building roof were approximately 400 mrem per hour "
at 18 in.
After recognizing that the high radiation area was not posted appropriately, the radiation
protection supervisor directed the technicians to post the area as a high radiation area.'
Later that evening, a radiation protection supervisor reviewed the posting again and
recognized that, while the area had been posted as a high radiation area, it had not been
barricaded. The supervisor barricaded the area immediately with yellow and maconta
rope. Later, radiation protection supervisors initiated PIRs 98-1027 and -1028 to address
these issues.
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Technical Specification 6.12 required this area to be posted as a high radiation area and
barricaded. The failure of technicians to properly post and barricade this area are two
examples of a violation of Technical Specification 6.12 (50-482/9810-11).
c. Conclusions
Radiation protection technicians improperly posted an area with dose rates of
approximately 400 mrem /hr at 18 in. After discovery, the technicians properly posted
but failed to barricade the area.
R1.2 Locked High Radiation Area Kev Issued Inanorooriatelv
a. Insoection Scooe (71750)
The inspector reviewed the licensee's actions following their discovery of an instance in
which a locked high radiation area key was issued without the approval of radiation
protection personnel.
b. Observations and Findings
On April 10,1998, security personnelissued the locked high radiation area keys to a
radwaste operator inappropriately in that they did not have the permission of radiation ;
protection p;rsonnel. Security Procedure 01-206, "High Security Key Control and Issue,"
Revision 23, required that locked high radiation area keys only be issued to individuals
listed on the authorized list kept at the secondary alarm station. Only radiation protection-
technicians and their supervisors were on the authorized list.
Security personnel recently reorganized their key storage. The locksmith provided a
reminder to all security sergeants and lieutenants that keys were being reorganized anci
that additional effort would be needed in the area of self-checking and attention to detail.
However, the sergeant who issued the key did not ask the radwaste operator which key
was needed and did not carefully check to ensure that the key issued was the one
intended.
Immediate corrective actions included contacting the radwaste operator, retrieving the
' locked high radiation area key, obtaining a voluntary statement from the radwaste
operator, notifying radiation protection personnel, and briefing all key issue officers
regarding the event and the importance of issuing keys properly. The security lieutenant
initiated PIR 98-1029 to address longer-term corrective actions. One longer-term
corrective action that was completed within a few days of the event was to attach yellow
and magenta tags to each locked high radiation area key to readily identify these keys.
While the radwaste operator entered the radiologically controlled area with the locked
high radiation area key for shift turnover, according to the voluntary statement, the key
was never used to access locked high radiation areas.
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The inappropriate issuance of a locked high radiation area key is a violation of Technica! l
Specification 6.11. This licensee-identified and corrected violation is being treated as a
noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy
(50-482/9810-12).
c. Conclusions )
Security personnel inadvertently issued a locked high radiation area key to a radwaste l
opcrator who was not authorized to receive this key. )
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R1.3 Unclear Contaminated Area Postings
I
a. Insoection Scooe (71750) )
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Throughout this inspection period, during tours in the radiologically controlled area, the
inspectors observed various radiation protection postings.
b. Observations and Findinas l
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On April 3,1998, the inspector noted contaminated area postings for l
Valves EJ HV8816A and -B, the residual heat removal discharge cross tie to reactor j
coolant system hot leg injection. These postings warned that there was a contaminated '
area "inside valve body." The inspector asked the lead health physics technician what
these postings meant and was informed that there was a packing leak and that the
contaminated area was on the valve body in the valve stem packing area.
The lead radiation protection technician subsequently reposted the valves to more clearly j
describe the proper location of the contaminated area. Radiation protection supervision
also stated that this issue would be reinforced during future discussions with all radiation ]
protection technicians. J
c. Conclusions ;
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Two contaminated area postings did not clearly describe the location of contamination on i
valves.
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R3 Radiological Protection and Chemistry Procedures and Documentation !
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R3.1 Radiation orotection ALARA Review of Irradiated Soecimen Transfer Procedure
a. Insoection Scooe (71750)
The inspector reviewed radiation protection department activities associated with transfer
of an irradiated specimen from the spent fuel pool to the shipping cask.
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b. Observations and Findinas
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On March 25,1998, the inspector observed engineering, operations, and radiation
protection personnel prepare for and attempt to accomplish movement of an irradiated
reactor vessel specimen from the spent fuel po% w here it had been stored since
Refueling Outage 9, to a shipping cask along sie the pool. The radiological controls I
were established in Radiation Work Permit 980087, Revision 1.
The irradiated specimen transfer process had been accomplished twice before this time,
(nost recently in December of 1991. This infrequent task involved raising the irradiated
specimen from the spent fuel pool using the spent fuel pool crane and lowering the
specimen into a shipping cask with a narrow (just-the-right-size) opening. This process
_
was complicated by the fact that high radiation doses from the specimen (40R/hr)
required that operators and radiation protection support personnel perform their function
from a distance in order to prevent high doses and to reduce their radiation doses.
Procedure AP 16C-003, " Work Package Task Planning," Revision 5, paragraph 6.5.2.11,
refers the work package planner to Procedure AP 25B-300, "RWP Program," Revision 6,
to use a " Pre-Job ALARA Checklist" if the work was to be accomplished in the
radiological controlled area. Procedure AP 258-300, paragraph 5.4, requires that prejob
ALARA checklists be performed only when estimated doses were expected to be equal
or greater than one person-rem. The estimated doses for the irradiated specimen
transfer were less than one person-rem, though the contingency plan in the event that
the specimen became stuck while being inserted into the shipping cask included sending
an individual up on a ladder to grasp the specimen by hand to physically assist inserting
the specimen into the cask. If the this planned contingent action had been called upon,
the exposure for the entire specimen transfer could have exceeded the one person-rem
criteria for an ALARA review. Despite this contingency plan and the potential for high
doses should a problem have been encountered, no ALARA review per
Procedure AP 258-300 was performed.
'The inspectors found that the level of ALARA review of the work package and the
procedure used to transfer the irradiated specimen was not commensurate with the
potential risk to workers involved in the specimen transfer process. The inspectors
determined that this was partly a result of Procedure AP 25B-300 not including potential
exposure in the criteria for determining when an ALARA review should be performed.
The inspectors noted that the radiation protection department started reviewing these
procedures and policy for possible correction.
c. Conclusions
The inspectors found that the level of ALA RA review of an irradiated specimen transfer
work package and procedure was not con mensurate with the potential risk to workers
involved in the specimen transfer process. The inspectors determined that this was
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because the licensee's radiation protection program did not include the potential for high
exposure in the criteria for determining when an ALARA review should be performed
(Section 3.1).
V. Manaaement Meetinas
X1 Exit Meeting Summary
The inspectors presented the inspection results to members of licensee management at the '
conclusions of the inspection on April 17,1998. The licensee acknowledged the findings l
presented.
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The inspectors asked the licensee whether any materials examined during the inspection should I
be considered proprietary. No proprietary information was identified.
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SUPPLEMENTAL INFORMATION
PARTIAL LIST OF PERSONS CONTACTED
Licensee
M. J. Angus, Manager, Licensing and Corrective Action
G. D. Boyer, Chief Administrative Officer
J. W. Johnson, Manager, Resource Protection
O. L. Maynard, President and Chief Executive Officer
B. T. McKinney, Plant Manager
R. Muench, Vice President Engineering
]
W. B. Norton, Manager, Performance improvement and Assessment 1
C. C. Warren, Chief Operating Officer )
1
INSPECTION PROCEDURES USED
IP 37551 Onsite Engineering
IP 61726 Surveillance Observations
IP 62707 Maintenance Observations
IP 71707 Plant Operations
IP 71750 Plant Support Activities
IP 92901 Followup - Operations
ITEMS OPENED AND CLOSED
Ooened
50-482/9810-01 NCV Failure to comply with Technical Specification Surveillance
Requirement 4.3.2.2 (Section 08.1)
1
50-482/9810-02 NCV Failure to comply with Technical Specification Action 1
Statement 3.6.1.7.a (Section 08.2)
50-482/9810-03 NCV Failure to comply with surveillance requirement to test
certain components at shutdown (Section 08.3)
50-482/9810-04 NCV Failure to accurately perform the containment integrated
leak rate test (Section 08.4)
50-482/9810-05 NCV Violation of Technical Specification 3./8.1.1., Emergency
Diesel Generator B (Section 08.5) l
50-482/9810-06 NCV Failure to procure test gas for a safety-related component
through an Appendix B program (Section 08.6)
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50-482/9810-07 ~ NCV Section 3.7.3 Technical Specification violation regarding two
independent component cooling water loops (Section 08.8) 'l
50-482/9810-08 NCV Waste gas system hydrogen / oxygen analyzers not operated
in accordance with Technical Specification requirements
(Section 08.9)
50-482/9810-09 NCV Failure to adequately vent emergency core cooling pump
casings and discharge piping high points in accordance with
Technical Specification Surveillance Requirement 4.5.2b.1
(Section 08.10)
50-482/9810-10 URI Molded case breaker are flash event (Section M3.1)
50-482/9810-11 VIO Inadequate high radiation area posting (Section R1.1)
50-482/9810-12 NCV Locked high radiation area key issued
inappropriately (Section R1.2)
Closed
50-482/9618-00 LER Failure to comply with Technical Specification Surveillance
Requirement 4.3.2.2 (Section 08.1)
50 482/9620-00 LER Failure to comply with Technical Specification Action
Statement 3.6.1.7.a (Section 08.2)
50-482/9701-00, LER Failure to comply with surveillance requirement to test certain
01,02,03 components at shutdown (Section 08.3) -
50-482/9704-00 LER Failure to accurately perform the containment integrated leak
rate test (Section 08.4)
50-482/9705-00 LER Violation of Technical Specification 3.8.1.1, Emergency
Diesel Generator B (Section 08.5)
- 50-482/9707-00, LER. Failure to procure test gas for a safety-related component
.01- through an Appendix B program (Section 08.6)
50-482/9709-01 LER Failure to comply with Technical Specification 4.5.2.c.2
(Section 08.7)
50-482/9713-00 LER Section 3.7.3 Technical Specification violation regarding two
independent component cooling water loops (Section 08.8)
- 50-482/9714-00 LER Waste gas system hydrogen / oxygen analyzers not operated
in accordance with Technical Specification requirements
(Section 08.9)
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50-482/9719-00 LER Failure to adequately vent emergency core cooling pump
casings and discharge piping high points in accordance with
Technical Specification Surveillance Requirement 4.5.2b.1
(Section 08.10)
50-482/9810-01 NCV Failure to comply with Technical Specification Surveillance
Requirement 4.3.2.2 (Section 08.1)
50-482/9810-02 NCV Failure to comply with Technical Specification Action
Statement 3.6.1.7.a (Section 08.2)
50-482/9810-03 NCV Failure to comply with surveillance requirement to test certain
components at shutdown (Section 08.3)
50-482/9810-04 NCV Failure to accurately perform the containment integrated leak
rate test (Section 08.4)
50-482/9810-05 NCV Violation of Technical Specification 3.8.1.1, Emergency
Diesel Generator B (Section 08.5)
50-482/9810-06 .NCV Failure to procure test gas for a safety-related component
through an Appendix B program (Section 08.6)
50-482/9810-07 NCV Section 3.7.3 Technical Specification violation regarding two
independent component cooling water loops (Section O8.8)
50-482/9810-08 NCk Waste gas system hydrogen / oxygen analyzers not operated
in accordance with Technical Specification requirements
(Section 08.9)
50-482/9810-09 NCV Failure to adequately vent emergency core cooling pump
casings and discharge piping high points in accordance with
Technical Specification Surveillance Requirement 4.5.2b.1
(Section 08.10)
50-482/9810-12 NCV Locked high radiation area key issued
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inappropr'ately (Section R1.2)
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